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Fission energy release #197

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e0f0e44
Added fission * energy release as a cross section
ChasingNeutrons Feb 24, 2026
66f39c5
Merge branch 'develop' into kappa_fission
ChasingNeutrons Mar 3, 2026
c368354
Fixes
ChasingNeutrons Mar 3, 2026
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23 changes: 22 additions & 1 deletion NuclearData/Reactions/reactionMG/fissionMG_class.f90
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Expand Up @@ -18,6 +18,8 @@ module fissionMG_class
!!
public :: fissionMG_TptrCast

real(defReal), private, parameter :: KAPPA_DEFAULT = 202.27 ! [MeV]

!!
!! A special type of MG reaction that contains data related to fission
!!
Expand All @@ -26,14 +28,17 @@ module fissionMG_class
!! Fission spectrum is independent of the energy group
!!
!! Public members:
!! data -> data for fissionMG [energyGroup, reaction (nu or chi)]
!! data -> data for fissionMG [energyGroup, reaction (nu or chi)]
!! kappa -> Energy per fission. 200 MeV by default.
!!
!! Interface:
!! reactionMG interface
!! buildFromDict -> builds fissionMG from a SCONE dictionary
!! getKappa -> obtains the energy per fission in MeV
!!
type, public, extends(reactionMG) :: fissionMG
real(defReal),dimension(:,:),allocatable :: data
real(defReal) :: kappa = KAPPA_DEFAULT
contains
! Superclass procedures
procedure :: init
Expand All @@ -46,6 +51,7 @@ module fissionMG_class

! Local procedures
procedure :: buildFromDict
procedure :: getKappa

end type fissionMG

Expand Down Expand Up @@ -216,6 +222,9 @@ subroutine buildFromDict(self, dict)

! Get number of groups
call dict % get(nG, 'numberOfGroups')

! Get energy per fission
call dict % getOrDefault(self % kappa, 'kappa', KAPPA_DEFAULT)

! Allocate space
allocate(self % data(nG, 2))
Expand Down Expand Up @@ -246,6 +255,18 @@ subroutine buildFromDict(self, dict)

end subroutine buildFromDict

!!
!! Get the energy release from fission
!!
pure function getKappa(self) result(kappa)
class(fissionMG), intent(in) :: self
real(defReal) :: kappa

kappa = self % kappa

end function getKappa


!!
!! Cast reactionHandle pointer to fissionMG pointer
!!
Expand Down
17 changes: 17 additions & 0 deletions NuclearData/Reactions/uncorrelatedReactionCE/fissionCE_class.f90
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Expand Up @@ -54,6 +54,7 @@ module fissionCE_class
!! eLawPrompt -> Energy Law for the prompt fission neutrons
!! nuBarDelayed -> Delayed release table for incindent energy [MeV]
!! delayed -> Information about Delayed emission precursors
!! Q -> Q value for the reaction [MeV]
!!
!! Interface:
!! uncorrelatedReactionCE interface
Expand All @@ -65,6 +66,7 @@ module fissionCE_class
class(energyLawENDF), allocatable :: eLawPrompt
class(releaseLawENDF),allocatable :: nuBarDelayed
type(precursor),dimension(:),allocatable :: delayed
real(defReal) :: Q = ZERO

contains
! Superclass procedures
Expand All @@ -79,6 +81,7 @@ module fissionCE_class

! Type specific procedures
procedure :: buildFromACE
procedure :: getQ
end type fissionCE

contains
Expand Down Expand Up @@ -148,8 +151,21 @@ elemental subroutine kill(self)
deallocate(self % delayed)
end if

self % Q = ZERO

end subroutine kill

!!
!! Returns the Q-value
!!
pure function getQ(self) result(Q)
class(fissionCE), intent(in) :: self
real(defReal) :: Q

Q = self % Q

end function getQ

!!
!! Returns true if reaction is in Centre-Of-Mass frame
!!
Expand Down Expand Up @@ -349,6 +365,7 @@ subroutine buildFromACE(self, ACE, MT)
! Read basic data
call new_totalNU(self % nuBarTotal, ACE)
call new_energyLawENDF(self % eLawPrompt, ACE, MT)
self % Q = ACE % QforMT(MT)

! Read Delayed Data
if (withDelayed) then
Expand Down
3 changes: 3 additions & 0 deletions NuclearData/ceNeutronData/aceDatabase/aceNeutronDatabase_class.f90
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Expand Up @@ -937,6 +937,9 @@ subroutine init(self, dict, ptr, silent )

end if

! Obtain the energy per fission with which to scale fission heating
call dict % getOrDefault(self % H235, 'energyPerFission', 202.27_defReal)

! Get path to ACE library
call dict % get(aceLibPath,'aceLibrary')

Expand Down
53 changes: 37 additions & 16 deletions NuclearData/ceNeutronData/aceDatabase/aceNeutronNuclide_class.f90
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Expand Up @@ -41,14 +41,17 @@ module aceNeutronNuclide_class


! Grid location parameters
integer(shortInt), parameter :: TOTAL_XS = 1
integer(shortInt), parameter :: ESCATTER_XS = 2
integer(shortInt), parameter :: IESCATTER_XS = 3
integer(shortInt), parameter :: CAPTURE_XS = 4
integer(shortInt), parameter :: FISSION_XS = 5
integer(shortInt), parameter :: NU_FISSION = 6
integer(shortInt), parameter :: PROMPT_NU_FISSION = 7
integer(shortInt), parameter :: TOTAL_XS = 1
integer(shortInt), parameter :: ESCATTER_XS = 2
integer(shortInt), parameter :: IESCATTER_XS = 3
integer(shortInt), parameter :: CAPTURE_XS = 4
integer(shortInt), parameter :: FISSION_XS = 5
integer(shortInt), parameter :: NU_FISSION = 6
integer(shortInt), parameter :: KAPPA_XS = 7
integer(shortInt), parameter :: PROMPT_NU_FISSION = 8

integer(shortInt), parameter :: NON_FISSILE_SIZE = 4, &
FISSILE_SIZE = 8

!!
!! Groups data related to an MT reaction
Expand Down Expand Up @@ -437,10 +440,12 @@ elemental subroutine microXSs(self, xss, idx, f)
if (self % isFissile()) then
xss % fission = data(FISSION_XS, 2) * f + (ONE-f) * data(FISSION_XS, 1)
xss % nuFission = data(NU_FISSION, 2) * f + (ONE-f) * data(NU_FISSION, 1)
xss % kappaXS = data(KAPPA_XS, 2) * f + (ONE-f) * data(KAPPA_XS, 1)
xss % promptNuFission = data(PROMPT_NU_FISSION, 2) * f + (ONE-f) * data(PROMPT_NU_FISSION, 1)
else
xss % fission = ZERO
xss % nuFission = ZERO
xss % kappaXS = ZERO
xss % promptNuFission = ZERO
end if
end associate
Expand Down Expand Up @@ -486,10 +491,12 @@ subroutine getThXSs(self, xss, idx, f, E, kT, rand)
if (self % isFissile()) then
xss % fission = data(FISSION_XS, 2) * f + (ONE-f) * data(FISSION_XS, 1)
xss % nuFission = data(NU_FISSION, 2) * f + (ONE-f) * data(NU_FISSION, 1)
xss % kappaXS = data(KAPPA_XS, 2) * f + (ONE-f) * data(KAPPA_XS, 1)
xss % promptNuFission = data(PROMPT_NU_FISSION, 2) * f + (ONE-f) * data(PROMPT_NU_FISSION, 1)
else
xss % fission = ZERO
xss % nuFission = ZERO
xss % fission = ZERO
xss % nuFission = ZERO
xss % kappaXS = ZERO
xss % promptNuFission = ZERO
end if

Expand Down Expand Up @@ -560,10 +567,12 @@ subroutine getUrrXSs(self, xss, idx, f, E, xi)
if (self % isFissile()) then
xss % fission = data(FISSION_XS, 2) * f + (ONE-f) * data(FISSION_XS, 1)
xss % nuFission = data(NU_FISSION, 2) * f + (ONE-f) * data(NU_FISSION, 1)
xss % kappaXS = data(KAPPA_XS, 2) * f + (ONE-f) * data(KAPPA_XS, 1)
xss % promptNuFission = data(PROMPT_NU_FISSION, 2) * f + (ONE-f) * data(PROMPT_NU_FISSION, 1)
else
xss % fission = ZERO
xss % nuFission = ZERO
xss % fission = ZERO
xss % nuFission = ZERO
xss % kappaXS = ZERO
xss % promptNuFission = ZERO
end if

Expand Down Expand Up @@ -594,6 +603,7 @@ subroutine getUrrXSs(self, xss, idx, f, E, xi)

if (self % isFissile()) then
xss % nuFission = xss % nuFission/xss % fission * val(3)
xss % kappaXS = xss % kappaXS / xss % fission * val(3)
xss % promptNuFission = xss % promptNuFission/xss % fission * val(3)
xss % fission = val(3)
end if
Expand Down Expand Up @@ -733,6 +743,7 @@ subroutine init(self, ACE, nucIdx, database)
top, firstIdxMT4
real(defReal), dimension(:), allocatable :: xsMT4
type(stackInt) :: scatterMT, absMT
real(defReal) :: H_Q
character(100), parameter :: Here = "init (aceNeutronNuclide_class.f90)"

! Reset nuclide just in case
Expand All @@ -753,9 +764,9 @@ subroutine init(self, ACE, nucIdx, database)

! Allocate space for main XSs
if (self % isFissile()) then
N = 7
N = FISSILE_SIZE
else
N = 4
N = NON_FISSILE_SIZE
end if

allocate(self % mainData(N, Ngrid))
Expand Down Expand Up @@ -815,10 +826,20 @@ subroutine init(self, ACE, nucIdx, database)
call self % fission % init(ACE, N_f)
end if

! Calculate nuFission
! Obtain Heating/Q scaling ratio
! Check if database is associated in order to satisfy tests where it might not be!
if (associated(database)) then
H_Q = database % H235 / database % Q235
else
H_Q = ONE
end if

! Calculate nuFission and kappaXS
do i = bottom, Ngrid
self % mainData(NU_FISSION,i) = self % mainData(FISSION_XS,i) * &
self % fission % release(self % eGrid(i))
self % mainData(NU_FISSION,i) = self % mainData(FISSION_XS,i) * &
self % fission % release(self % eGrid(i))
self % mainData(KAPPA_XS,i) = self % mainData(FISSION_XS,i) * &
self % fission % getQ() * H_Q
self % mainData(PROMPT_NU_FISSION,i) = self % mainData(FISSION_XS,i) * &
self % fission % releasePrompt(self % eGrid(i))
end do
Expand Down
6 changes: 6 additions & 0 deletions NuclearData/ceNeutronData/ceNeutronDatabase_inter.f90
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Expand Up @@ -39,6 +39,8 @@ module ceNeutronDatabase_inter
!!
!! Public Members:
!! mapDBRCnuc -> map to link indexes of DBRC nuclides with their corresponding 0K
!! H235 -> heating fission of U235, used for scaling fission energy
!! Q235 -> Q-value of U235 fission, used for scaling fission energy
!!
!! Interface:
!! nuclearDatabase Interface
Expand All @@ -53,6 +55,10 @@ module ceNeutronDatabase_inter
!!
type, public, abstract, extends(nuclearDatabase) :: ceNeutronDatabase
type(intMap) :: mapDBRCnuc

! Default fission scaling [MeV]
real(defReal) :: H235 = 202.27_defReal
real(defReal) :: Q235 = 193.406_defReal

contains

Expand Down
2 changes: 2 additions & 0 deletions NuclearData/mgNeutronData/baseMgNeutron/baseMgNeutronMaterial_class.f90
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Expand Up @@ -129,9 +129,11 @@ subroutine getMacroXSs_byG(self, xss, G, rand)
if(self % isFissile()) then
xss % fission = self % data(FISSION_XS, G)
xss % nuFission = self % data(NU_FISSION, G)
xss % kappaXS = xss % fission * self % fission % getKappa()
else
xss % fission = ZERO
xss % nuFission = ZERO
xss % kappaXS = ZERO
end if

end subroutine getMacroXSs_byG
Expand Down
11 changes: 10 additions & 1 deletion NuclearData/testNeutronData/testNeutronDatabase_class.f90
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Expand Up @@ -71,18 +71,20 @@ module testNeutronDatabase_class
!! captureXS [in] -> Optional. Value of Capture XS
!! fissionXS [in] -> Optional. Value of Fission XS
!! nuFissionXS [in] -> Optional Value of nuFission
!! kappaXS [in] -> Optional Value of kappa XS
!!
!! Errors:
!! None
!!
subroutine build(self, xsVal, eScatterXS, ieScatterXS ,captureXS, fissionXS, nuFissionXS)
subroutine build(self, xsVal, eScatterXS, ieScatterXS ,captureXS, fissionXS, nuFissionXS, kappaXS)
class(testNeutroNDatabase), intent(inout) :: self
real(defReal), intent(in) :: xsVal
real(defReal), intent(in),optional :: eScatterXS
real(defReal), intent(in),optional :: ieScatterXS
real(defReal), intent(in),optional :: captureXS
real(defReal), intent(in),optional :: fissionXS
real(defReal), intent(in),optional :: nuFissionXS
real(defReal), intent(in),optional :: kappaXS

self % xsVal = xsVal

Expand Down Expand Up @@ -126,6 +128,13 @@ subroutine build(self, xsVal, eScatterXS, ieScatterXS ,captureXS, fissionXS, nuF
else
self % mat % xss % nuFission = xsVal
end if

! kappa * fission
if(present(kappaXS)) then
self % mat % xss % kappaXS = kappaXS
else
self % mat % xss % kappaXS = xsVal
end if

end subroutine build

Expand Down
12 changes: 10 additions & 2 deletions NuclearData/xsPackages/neutronXsPackages_class.f90
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Expand Up @@ -2,7 +2,6 @@
!! This module brakes standard rules
!! It contains a library of XS Packages for Neutron particle type
!!
!!
module neutronXsPackages_class

use numPrecision
Expand All @@ -22,6 +21,7 @@ module neutronXsPackages_class
!! capture -> sum of all reactions without secendary neutrons excluding fission [1/cm]
!! fission -> total Fission MT=18 Cross-section [1/cm]
!! nuFission -> total average neutron production Cross-section [1/cm]
!! kappaXS -> heating cross-section [MeV/cm]
!! promptNuFission -> prompt-only average neutron production Cross-section [1/cm]
!!
!! Interface:
Expand All @@ -36,6 +36,7 @@ module neutronXsPackages_class
real(defReal) :: capture = ZERO
real(defReal) :: fission = ZERO
real(defReal) :: nuFission = ZERO
real(defReal) :: kappaXS = ZERO
real(defReal) :: promptNuFission = ZERO
contains
procedure :: clean => clean_neutronMacroXSs
Expand All @@ -56,6 +57,7 @@ module neutronXsPackages_class
!! capture -> all reactions without secendary neutrons excluding fission [barn]
!! fission -> total Fission MT=18 Cross-section [barn]
!! nuFission -> total average neutron production Cross-section [barn]
!! kappaXS -> heating Cross-section [MeV*barn]
!! promptNuFission -> prompt-only average neutron production Cross-section [barn]
!!
type, public :: neutronMicroXSs
Expand All @@ -65,6 +67,7 @@ module neutronXsPackages_class
real(defReal) :: capture = ZERO
real(defReal) :: fission = ZERO
real(defReal) :: nuFission = ZERO
real(defReal) :: kappaXS = ZERO
real(defReal) :: promptNuFission = ZERO
contains
procedure :: invert => invert_microXSs
Expand Down Expand Up @@ -92,6 +95,7 @@ elemental subroutine clean_neutronMacroXSs(self)
self % capture = ZERO
self % fission = ZERO
self % nuFission = ZERO
self % kappaXS = ZERO
self % promptNuFission = ZERO

end subroutine clean_neutronMacroXSs
Expand Down Expand Up @@ -119,6 +123,7 @@ elemental subroutine add_neutronMacroXSs(self, micro, dens)
self % capture = self % capture + dens * micro % capture
self % fission = self % fission + dens * micro % fission
self % nuFission = self % nuFission + dens * micro % nuFission
self % kappaXS = self % kappaXS + dens * micro % kappaXS
self % promptNuFission = self % promptNuFission + dens * micro % promptNuFission

end subroutine add_neutronMacroXSs
Expand Down Expand Up @@ -165,6 +170,9 @@ elemental function get(self, MT) result(xs)
case(macroNuFission)
xs = self % nuFission

case(macroKappaFission)
xs = self % kappaXS

case(macroPromptNuFission)
xs = self % promptNuFission

Expand Down Expand Up @@ -245,7 +253,7 @@ end function invert_macroXSs
!!
!! Use a real r in <0;1> to sample reaction from Microscopic XSs
!!
!! This function involves a bit of code so is written for conviniance
!! This function involves a bit of code so is written for convenience
!!
!! Args:
!! r [in] -> Real number in <0;1>
Expand Down
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